Core materials for fast reactors are required to have excellent resistance to creep deformation and environment-induced cracking at high temperatures in high-radiation environments. In a prototype reactor Monju, used is an SUS316 austenitic stainless steel for the fuel-cladding pipes to which the most stringent conditions are applied among core materials for fast reactors.
As the abrasion-resistant high-strength members of light-water reactors, used is Stellite or Inconel; however, regarding Stellite, radioactivation associated with the Co-based alloy is an important issue in maintenance, and regarding Inconel, grain-boundary stress corrosion cracking is an important issue. As a candidate of a material for fuel-cladding pipes of supercritical water-cooled reactors, a SUS310 steel is exemplified; however, on a high-temperature side at 700° C., the austenite phase stability thereof is low, and therefore α-phase brittlement is an important issue.
FIG. 9 shows a relationship between TTC (aging time-temperature-corrosion range) diagrams relating to aging embrittlement of an SUS316 steel austenite (γ phase) and the γ phase stability thereof. In the left-hand chart, IGC is grain-boundary corrosion relating to the formation of a σ phase and a Cr-depleted layer of carbide in a grain boundary, and MPC (Martensite Path Corrosion) is a depression region of corrosion resistance at a grain boundary and within a grain, which is accompanied by the formation of a high-Cr plane and a low-Cr plane on a crystal plane level through spinodal decomposition of processing-induced martensite or remanent ferrite, which is generated depending on a heat hysteresis of cold working or the like and the aging time. This is because the γ phase itself is low in stability in a supercooled state and structure change is occurred during aging in a practical temperature range, as in the right-hand chart.
In a fast reactor of a high-temperature operation, a heat-resistant alloy resistant to high temperatures on a level of 700° C. is needed for the heat-transfer pipes of a heat exchanger; but existing Inconel 690 and the like are not precipitation-strengthened ones and the mechanical strength thereof is unsatisfactory.
Existing materials of austenitic stainless steel such as JIS SUS304 or SUS316 contain Cr in an amount of 16 wt % or more and may readily form a passivation film to exhibit excellent corrosion resistance, and in addition, the materials have excellent forming-workability and toughness intrinsic to face-centered cubic crystals, and are widely used as core structure materials for atomic reactors. However, these materials have a low Ni content of 20% or less, and therefore the thermodynamic stability of the austenite phase itself at operating temperatures is insufficient. Under the conditions of fast reactors that receive heavy irradiation by fast neutrons in a broad temperature range of from 250 to 700° C., the irradiation embrittlement is easily occurred in a low- or middle-temperature range of from 250 to 450° C., and deformation is easily occurred owing to volume change and irradiation creep by void swelling at 450° C. or higher. Consequently, the materials are insufficient in point of the heat resistance and the irradiation resistance, and are difficult to be applied to fast reactors.
FIG. 10 shows the influence of γ phase stability on irradiation resistance (void swelling resistance). The irradiation resistance is most excellent under the condition of Fe/Ni ratio at which γ phase is stabilized as a solid solution.
As alloys having improved irradiation resistance, austenitic stainless steels such as PNC316 and PNC1520 have been developed. By using these, the latent period before void swelling is controlled. However, after void swelling is occurred, the growth/coarsening of void occurs in proportion to the irradiance level (time), and therefore prototype reactors on an irradiance level of up to about 100 dpa are in the applicable range, and any other material more excellent in irradiation resistance is needed for practical reactors on an irradiance level of 250 dpa.
Given the situation, as a fuel-cladding pipe material for future fast reactors capable of solving the problems with existing austenitic stainless steels, studies and developments of precipitation-strengthened Ni-based heat resistant alloys are being made actively in Europe and America; and Nimonic Alloy PE16 used in prototype reactors in England, and new alloys created by modifying US Inconel-type commercial alloys have been developed. Patent Reference 1 discloses an Fe—Ni-based austenitic alloy excellent in neutron irradiation resistance and sodium corrosion resistance, which is used for core members of fast-breeder reactors such as fuel-cladding pipes.
As a technology of materials in another system for solving the problems with existing austenitic stainless steels, for example, a ferritic steel has been developed. A ferritic steel has a body-centered cubic crystal that hardly experience void swelling, and therefore mainly in US, HT9 or the like is being used for metal fuel-cladding pipes of fast reactors for use as breeder reactors operating at low temperatures. However, the mechanical strength at high temperature of ferritic steel is low as compared with that of austenitic steel; and therefore, the heat resistance thereof is problematic. Accordingly, Patent Reference 2 discloses a martensitic oxide dispersion steel (martensitic ODS steel) excellent in high-temperature strength.